ML21200A179

Scale/Melcor Non-LWR Source Term Demonstration Project - High Temperature Gas-Cooled Reactor

ML21200A179
Person / Time
Site:	Grand Gulf, Arkansas Nuclear, River Bend, Waterford, 07000013
Issue date:	07/31/2021
From:	Jordan Hoellman NRC/NRR/DANU/UARP, Oak Ridge, Sandia
To:
Hoellman J
References
DE-NA0003525
Download: ML21200A179 (115)

Text

SCALE/MELCOR Non-LWR Source Term Demonstration Project -

High-Temperature Gas-Cooled Reactor July 2021 Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energys National Nuclear Security Administration under contract DE-NA0003525. SAND20XX-XXXX P

Outline NRC strategy for non-LWR source term analysis Project scope High-temperature gas-cooled reactor fission product inventory/decay heat methods and results High-temperature gas-cooled reactor plant model and source term analysis Summary Appendices

• SCALE overview

• VSOP

• ORIGEN library interpolation

• MELCOR overview

• MELCOR default radionuclide classes 2

Integrated Action Plan (IAP) for Advanced Reactors Near-Term Implementation Action Plan Strategy 1 Strategy 4 Knowledge, Skills, Industry Codes and Capacity and Standards Strategy 5 Strategy 2 Technology Analytical Tools Inclusive Issues ML17165A069 Strategy 3 Strategy 6 Flexible Review Communication Process 3

IAP Strategy 2 Volumes These Volumes outline the specific analytical tools to enable independent analysis of non-LWRs, gaps in code capabilities and data, V&V needs and code development tasks.

Introduction Volume 1 ML20030A174 ML20030A176 Volume 3 Volume 2 Volume 4 Volume 5 ML20030A177 ML21085A484 ML21088A047 ML20030A178 4

NRC strategy for non-LWR analysis (Volume 3) 5

Role of NRC severe accident codes 6

Project Scope Project objectives Understand severe accident behavior

• Provide insights for regulatory guidance Facilitate dialogue on staffs approach for source term Demonstrate use of SCALE and MELCOR

• Identify accident characteristics and uncertainties affecting source term

• Develop publicly available input models for representative designs 8

Project scope Full-plant models for three representative non-LWRs (FY21)

• Heat pipe reactor - INL Design A

• Pebble-bed gas-cooled reactor - PBMR-400

• Pebble-bed molten-salt-cooled - UC Berkeley Mark I FY22

• Molten-salt-fueled reactor - MSRE

• Sodium-cooled fast reactor - To be determined 9

Project approach

1. Build MELCOR full-plant input model

• Use SCALE to provide decay heat and core radionuclide inventory

2. Scenario selection

3. Perform simulations for the selected scenario and debug

• Base case

• Sensitivity cases 10

Advanced Reactor Designs Broad Landscape Liquid Metal Cooled Fast High-Temperature Gas-Cooled Molten Salt Reactors Micro Reactors Reactors (MSR) Reactors (LMFR) (HTGR)

TerraPower/GEH (Natrium)* Kairos

• Westinghouse (eVinci)

X-energy

• GEH PRISM (VTR) Kairos (HermeslRTR) BWX Technologies Framatome Oklo Liquid Salt Cooled X-energy StarCore Advanced Reactor Radiant lRTR Concepts MIT Sodium-Cooled Terrestrial

• Transportable TerraPower Westinghouse TRISO Fuel Ultra Safe lRTR Southern (TP MCFR) lRTR Columbia Basin General Atomics (EM2) ACU lRTR

• Oklo Hydromine General Atomics Elysium Stationary Lead-Cooled Thorcon LEGEND ARDP Awardees Muons Demo In Licensing Review Flibe Reactors Risk Reduction

• Preapplication Alpha Tech ARC-20 RTR Research/Test Reactor Liquid Salt Fueled 11

High-Temperature Gas-Cooled Reactor

High-temperature gas-cooled reactor (1/2)

High-temperature high-pressure helium transfers heat from core to the secondary system

• Core outlet temperatures to 1000

• High temperature increases efficiency

• Fuel in a prismatic or a pebble bed core Peach Bottom Unit 1

• Operated 1966-1974

• 115 MW thermal power

• 37% efficiency, 88% availability Fort St. Vrain

• Operated 1979-1989

• 842 MW thermal power Peach Bottom Unit #1

[https://commons.wikimedia.org/wiki/File:Peach_Bottom_-Aerial_View_1.jpg 13

High-temperature gas-cooled reactor (2/2)

Department of Energy funded design of the Next Generation Nuclear Plant

• Project as established by Energy Policy Act of 2005

• Project started in 2007

• Initial focus on the PBMR-400 design

• Pebble Bed Modular Reactor (Pty) Ltd

• Focus of an Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) neutronics benchmark study [NEA/NSC/DOC(2013)10] PBMR-400

[NEA/NSC/DOC(2013)10]

• Areva SC-HTGR design was selected in 2012

• Department of Energy subsequently ended support 14

Publicly available design PBMR-400 - Used for SCALE/MELCOR demonstration project MELCOR model based on data from OECD/NEA neutronics benchmark project

• Development of MELCOR Input Techniques for High Temperature Gas-cooled Reactor Analysis, James Corson, Masters thesis, Texas A&M University, 2010 No description of confinement or secondary system

• MELCOR confinement model based on NGNP schematics

• Simplified secondary system used to estimate steady-state conditions 15

PBMR-400 (1/2) 400 MWt Helium coolant

• Pressure - 9 MPa (1300 psi)

• Core inlet - 500

• Core outlet - 900

• Core flowrate (downward) - 192 kg/s 452,000 TRISO pebbles in an annular core

• Core inner diameter - 2.0 m

• Core outer diameter - 3.7 m

• Core height - 11 m 92 GWD/MTU target burn-up Steel vessel with graphite reflectors Pieter J Venter, Mark N Mitchell, Fred Fortier, PBMR Reactor Design and Development, 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18), Beijing, China, August 7-12, 2005, SMiRT18- S02-2 16

PBMR-400 (2/2)

TRISO particle

• TRISO is a portmanteau for tristructural isotropic

• Kernel - 1.5 g U; 250 µm radius

• Porous carbon buffer layer TRISO particle

[INL/EXT-08-14497]

• 3 coatings to contain fission products TRISO pebble

• Contains 14,500 TRISO particles

• 25 mm radius

• 5 mm graphite outer shell TRISO pebble

[1]

17

HTGR Fission Product Inventory / Decay Heat Methods & Results

PBMR-400 benchmark used to represent PBR concepts Design features

• Fueled by graphite pebbles composed of UO2-bearing TRISO fuel particles (5-10% 235U)

• Pebbles circulate multiple passes through the core to high discharge burnup (~90 GWd/MTIHM)

Two cases evaluated

• Startup core: 1/3 fuel pebbles, 2/3 graphite dummy pebbles

• Equilibrium core: 110 material zones with pre-specified material compositions (100% fuel)

References:

1. Status and Prospects for Gas Cooled Reactor Fuels, IAEA-TECDOC-CD-1614, April 2009

2. OECD/NEA, PBMR Coupled Neutronics / Thermal-hydraulics Transient Benchmark I: The PBMR-400 PBMR-400 SCALE geometry Core Design, NEA/NSC/DOC(2013)10, July 2010 (S. Skutnik, ORNL) 19

Prior SCALE validation for HTGR systems (1/2)

• HTR-10 initial core critical benchmark

• Based on International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE) benchmark for HTR-10 initial core

• Graphite-coated, spherical fuel elements with TRISO fuel particles 3 cm fuel spheres at 17% 235U enrichment

• SCALE 6.0 with ENDF/B-VIII.0 nuclear data

• Figure of merit: System k-eigenvalue (keff)

SCALE consistent with MCNP to within -73+/-34 pcm MCNP and SCALE calculations both showed a moderate positive reactivity bias (1.4 +/- 0.4)%

Image: NUREG/CR-7107 G. Ilas, D. Ilas, R. P. Kelly, and E. E. Sunny, Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis, NUREG/CR-7107(ORNL/TM-2011/161), Jul. 2012 20

Prior SCALE validation for HTGR systems (2/2)

• HTR-PROTEUS critical benchmark

• IRPhE benchmark based upon critical experiments performed at PROTEUS facility (Paul Scherrer Institut, Switzerland) 10 deterministic pebble packing arrangements with 3 random close-packed arrangements Graphite-coated spherical fuel elements with TRISO fuel particles 3 cm radius graphite spheres (2.35 cm fuel region radius), 16.7% 235U enrichment

• Figure of merit: System k-eigenvalue (keff)

Difference with MCNP5 (pcm)

ENDF/B-VI ENDF/B-VII.0 Average Maximum Average Maximum Columnar hexagonal point-on-point (CHPOP) 422 +/- 93 667 +/- 82 804 +/- 87 1302 +/- 811 Hexagonal close-packed (HCP) 252 +/- 93 353 +/- 84 782 +/- 95 801 +/- 85 G. Ilas, D. Ilas, R. P. Kelly, and E. E. Sunny, Validation of SCALE for High Temperature Gas-Cooled Reactor Analysis, NUREG/CR-7107(ORNL/TM-2011/161), Jul. 2012 21

Workflow SCALE specific Generic End-user specific SCALE Inventory Other Binary Output Interface File MACCS Input SCALE Power SCALE Text distributions MELCOR Input Output Kinetics data

• SCALE capabilities used

• Relatively small amount of data except for

• KENO or Shift* 3D Monte Carlo transport nuclide inventory

• ENDF/B-VII.1 continuous energy physics

• new interface file developed for inventory using standard JSON format

• ORIGEN for depletion

• easily read in python and post-processed into

• Sequences MELCOR or MACCS input CSAS for reactivity (e.g. rod worth)

• contains nuclear data such as decay Q-value TRITON for reactor physics & depletion for traceability when performing UQ studies

• To be released with SCALE 6.3 22

General ORNL Methodology for Fuel Inventory

• ORNL has used a methodology with the Oak Ridge Isotope GENeration (ORIGEN) code to rapidly generate inventories using ORIGEN reactor libraries

• SCALE/ORIGEN use of fundamental nuclear data allows the following to be calculated from nuclide inventory (moles of each nuclide in a system)

• mass

• decay heat

• activity

• gamma emission

• neutron emissions

• With SCALE 6.2 (2016), the sequence ORIGAMI was released which is the modern approach of using ORIGEN reactor libraries 23

Plans for SCALE/ORIGAMI and HTGR Current Fuel Types

• Soon ORIGAMI will have a new PBMR-400 fuel type and the ability to generate (in seconds)

• fuel inventory for a PBMR-400 pebble

• initial enrichment

• specific power history

• cooling time

• Generalizing what we learn for the PBMR-400 will enable future HTGR fuel types

>50 different fuel types supported! 24

HTGR analysis with SCALE: Overview

• Key assumptions

• License applications will specify pebble circulation strategy and equilibrium core PBMR-400

• Analyzing the equilibrium core is the limiting case from equilibrium core an inventory/decay heat standpoint

• Main goals

• Evaluate neutronic characteristics

• Generate inventory and decay heat for the MELCOR nodalization which may differ from how the application specifies their equilibrium core isotopics

• Generate individual pebble inventory within a core zone/batch (e.g., difference between fresh vs. once-through pebble in a single core zone)

• Generate discharge pebble inventory/decay heat with sensitivity/uncertainty analysis 25

Analysis areas

1. Pebble packing

2. Temperature feedback PBMR-400

3. Radial/axial spectral variation equilibrium core

4. Pebble flow

5. TRITON model scope for ORIGEN library generation

• (i.e. what matters for producing one-group sections) 26

1. Pebble packing PBMR-400 benchmark specifies ~452,000 fuel pebbles with a packing fraction of 61%

Can be achieved using a BCC lattice (dodecahedral) of unbroken spheres, however substantial negative bias in keff observed due to local voids near reflector regions Present best estimate models use clipped pebbles at boundary to maintain uniform local packing fraction

• Similar to modeling approaches used for HTR-10 Image: S. Skutnik, ORNL J.-Y. Hong, S.-R. Wu, S.-C. Wu, D.-S. Chao, J.-H. Liang, Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre generated fuel composition library, Nuclear Engineering and Design, 374 (2021) 27

2. Temperature feedback (1/2)

Estimation of specific reactivity feedback components (e.g., temperature reactivity coefficients of fuel, moderator) requires detailed thermal hydraulic analysis of core Strong coupling between neutronics & thermal hydraulics Approach: Using system isotherms

• All system materials adjusted to a fixed temperature e.g., 300, 600, 900, 1200 K

• Does not afford specific isolation of moderator / fuel temperature coefficients PBMR-400 total neutron flux, from SCALE/Shift 3D Monte Carlo Calculation (S. Skutnik, ORNL) 28

2. Temperature feedback (2/2)

Strong temperature-driven spectral shifts, especially toward 239Pu low-lying resonance Fresh core Equilibrium core 29

3. Flux shape shows a top-weighted distribution due to pebble loading & depletion thermal flux Strong power peaking effects observed near graphite reflector regions (esp. interior) 30

3. Fast : thermal flux ratio (spectral index) sensitive to radial zone; relatively invariant axially Axial Radial Central regions Reflector-adjacent (outer)

Major spectral shifts primarily occur across radial zones; i.e., primarily need radial zone Origen libraries 31

4. Continuous circulation of pebbles in the core Approach: Equilibrium compositions derived from previous equilibrium core calculation with flowing pebbles (VSOP)

• Pebble locations currently treated as static in a full-core, 3-D Monte Carlo neutron transport calculation

• Discrete axial and radial material zones, representing spatially-dependent average at equilibrium after several months of operation Similarity to prior approaches:

• VSOP:1 Depletion of fixed core compositions to a pre-defined keff, then shuffle zones downward, reload pebbles at top of core and repeat. Depletion assumes admixture of fresh & burned pebbles exposed to same depleting flux

• HTR-10 multi-pass pebble burnup analysis2 follows similar procedure to VSOP

References:

1. HJ. Rütten, K.A. Haas, H. Brockmann, W. Scherer, V.S.O.P. (99/05) Computer Code System (2005)

PBMR-400 total neutron flux from

2. J.-Y. Hong, S.-R. Wu, S.-C. Wu, D.-S. Chao, J.-H. Liang, Burnup computations of multi-pass fuel loading SCALE/Shift 3D Monte Carlo scenarios in HTR-10 using a pre generated fuel composition library, Nuclear Engineering and Design, 374 (2021) calculation (S. Skutnik, ORNL) 32

4. Capturing possible pebble transit paths through the core (velocity differentials & cross-flow)

Current assumptions:

• Pebble transit dominated by vertical motion; can capture differential velocity across radial regions

• Active core modeled as a right-cylindrical annulus (cylindrical shell)

Similarity to prior approaches:

VSOP: Pebble transit assumed to be in parallel vertical dimensions unless user specifies otherwise HTR-10 burnup analysis normalizes pebble residence time based on assumed transit path (conical funneling); recycled pebbles uniformly redistributed across top of core J.-Y. Hong, S.-R. Wu, S.-C. Wu, D.-S. Chao, J.-H. Liang, Burnup computations of multi-pass fuel loading scenarios in HTR-10 using a pre generated fuel composition library, Nuclear Engineering and Design, 374 (2021) 33

5. ORIGEN library analysis strategy Evaluate PBMR-400 cross-sections & isotopic responses at different levels of model fidelity Lower fidelity High fidelity Lower computational cost High computational cost 34

5. ORIGEN library development:

reflected plane model

• Accounts for important radial effects

• Proximity to reflector

• Effects of nearest neighbor pebbles

• Can easily be tuned for different axial zones 35

5. Plane model captures important neighbor effects Plane Pebble 36

5. ORIGEN library generation based on 5 spectral zones

• Five separate cases constructed starting with a fresh pebble surrounded by non-depleting neighbors with compositions derived from PBMR-400 benchmark inventory ND-Set3

• Pebble depleted to discharge burnup surrounded by invariant neighbors rzone=1 rzone=2 rzone=3 rzone=4 rzone=5 37

5. Radial, temperature effects drive differences in 1-group XSs ORIGEN libraries 38

5. Radial zone effects far more prevalent than burnup effects for pebble bed depletion Outer regions Central region Spatial-driven differences in loss cross-sections relatively stable over burnup 39

5. Temperature (system isotherm) shows a large, region-dependent effect on 1G removal XS Outer regions Magnitude of XS differences due to radial location increases with system temperature Central region

• Gap between inner and outer regions grows with increasing temperature

• Implies a covariant relationship between location & temperature 40

Conclusions for pebble bed reactor ORIGEN library development

• Analysis areas

• For ORIGEN library generation

1. Pebble packing

• Burnup effects appear to be

2. Temperature feedback second-order, roughly linear in nature

3. Radial/axial spectral variation

• Radial distance from the reflector

4. Pebble flow is a first-order spectrum

5. TRITON model scope for characteristic ORIGEN library generation Must be accounted for in library generation Further details:

S. Skutnik, W. Wieselquist, Assessment of ORIGEN

• Temperature (system isotherm)

Reactor Library Development for Pebble-Bed also a first-order effect Reactors Based on the PBMR-400 Benchmark, Shows covariance with radial position ORNL/TM-2020/1886, July 2021 Available on osti.gov Driven primarily by graphite (reflector) temperature 41

MELCOR High-Temperature Gas-Cooled Reactor Model

MELCOR HTGR modeling Fission product release

• Release from TRISO kernel

• Radionuclide distributions within the layers in the TRISO particle and compact

• Release to coolant Other core models

• Graphite oxidation

• Intercell and intracell conduction

• Convection & flow

• Point kinetics

• Dust generation and resuspension 43

HTGR Components Fueled pebble core Unfueled pebble core

• Pebble Bed Reactor Fuel/Matrix Legend Components Fueled part of pebble TRISO Unfueled shell (matrix) is Fuel (FU)

GRAPHITE modeled as separate component Fuel radial temperature profile for GRAPHITE Matrix (MX) sphere Fluid B/C

• Prismatic Modular Reactor Fuel/Matrix Components Rod-like geometry Part of hex block associated with TRISO (FU) a fuel channel is matrix component Unfueled Sub-component model Fuel radial temperature profile for Fuel for zonal diffusion of cylinder Compact radionuclides through TRISO particle 44

Transient/Accident Solution Methodology Stage 0: Stage 1: Stage 2: Stage 3:

Normal Operation Normal Operation Normal Operation Accident Establish thermal state Diffusion Calculation Transport Calculation Diffusion & Transport calculation Time constant in HTGR Calculate steady state distribution of Calculate accident graphite structures is very Establish steady state radionuclides and graphite dust progression and radionuclide large distribution of throughout system (deposition on release radionuclides in TRISO surfaces, convection through flow Reduce heat capacities for particles and matrix paths) structures to reach steady state thermal conditions.

Reset heat capacities after steady state is achieved.

Elevation [m]

Temperature [K]

Representative reflector temperature response Example:

PBMR-400 Cs 2000 K Distribution in Primary System Time [min]

Temperature [K]

45

HTGR Radionuclide Diffusion Release Model Intact TRISO Particles

• One-dimensional finite volume diffusion equation solver for multiple zones (materials) Diffusivity Data Availability

• Temperature-dependent diffusion coefficients (Arrhenius form)

Porous Matrix TRISO Radionuclide UO2 UCO PyC SiC 1

=

+ = 0 Ag Some Some Carbon Graphite Extensive Some Overall Extensive Not investigated Cs Some Some Extensive Some Some Not found I Some Some Some Not found Not found Kr Some Some Not found Some Some Sr Some Some Extensive Some Some Xe Some Some Some Some Not found Data used in the demo calculation

[IAEA TECDOC-0978]

FP Species Kr Cs Sr Ag D (m2/s) Q D (m2/s) Q D (m2/s) Q D (m2/s) Q Intact TRISO Layer (J/mole) (J/mole) (J/mole) (J/mole)

Kernel (normal) 1.3E-12 126000.0 5.6-8 209000.0 2.2E-3 488000.0 6.75E-9 165000.0 Buffer 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 1.0E-8 0.0 PyC 2.9E-8 291000.0 6.3E-8 222000.0 2.3E-6 197000.0 5.3E-9 154000.0 SiC 3.7E+1 657000.0 7.2E-14 125000.0 1.25E-9 205000.0 3.6E-9 215000.0 Concentrations Matrix Carbon 6.0E-6 0.0 3.6E-4 189000.0 1.0E-2 303000.0 1.6E00 258000.0 Str. Carbon 6.0E-6 0.0 1.7E-6 149000.0 1.7E-2 268000.0 1.6E00 258000.0 Iodine assumed to behave like Kr CORSOR-Booth LWR scaling used to estimate other radionuclides 46

HTGR Radionuclide Release Models o Recent failures - particles failing within latest time-step (burst release, diffusion release in time-step) o Previous failures - particles failing on a previous time-step (time history of diffusion release) o Contamination and recoil Diffusion from intact TRISO Transfer to Released Distribution calculated failed to the TRISO matrix from diffusion model Release from TRISO failure Fuel FuelPebble Pebble Diffusion Released to the matrix Diffusion Failing recoil Intact recoil TRISO Intact TRISO Transition from Intact-to-failed recoil Failed Release from TRISO failed TRISO Contamination (Modified Booth) Failed Recoil fission source TRISO Diffusion 47

Graphite Oxidation Steam oxidation Reactions H2O or Air He Reactions Air oxidation Both steam and air include rate limit due to steam/air diffusion towards active oxidation surface ROX is the rate term in the parabolic oxidation equation [1/s]

48

COR Intercell Conduction Effective conductivity prescription for Dp=.06 m pebble bed (bed conductance) Kf=.154 W/m-K Ks = 26 W/m-K

• Zehner-Schlunder-Bauer with Breitbach-Barthels modification to the radiation term Effective conductivity prescription for prismatic (continuous solid with pores)

• Tanaka and Chisaka expression for effective radial conductivity (of a single PMR hex block)

Ks = 26 W/m-K Kf=.154 W/m-K

• A radiation term is incorporated in parallel with the pore conductivity

• Thermal resistance of helium gaps between hex block fuel elements is added in parallel via a gap conductance term 49

Interface Between Thermal-hydraulics and Pebble Bed Reactor Core Structures Heat transfer coefficient (Nusselt number) correlations for pebble bed convection:

• Isolated, spherical particles

• Use Tfilm to evaluate non-dimensional numbers, use maximum of forced and free Nu 4 3 3

= 2.0 + 0.6 1 1 = 2.0 + 0.6 1 2 1

• Constants and exponents accessible by sensitivity coefficient Flow resistance

• Packed bed pressure drop Loss coefficient relative to Ergun (original) coefficient at Re=1000 1 4

, = 1 + 2 1

+ 3 1

50

Point kinetics modeling Standard treatment 6

= + + 0

=1

= , = 1 6 Feedback models

• User-specified external input

• Doppler

• Fuel and moderator density 51

High-Temperature Gas-Cooled Reactor Plant Model and Source Term Analysis

Reactor vessel and core 176 186 166 126 136 146 156 6 170 115 125 135 145 155 165 170 304 125 135 145 155 165 124 134 144 154 164 114 124 134 144 154 164 Inlet Riser 123 133 143 153 163 303 Pebble bed Rings 2-6, 123 133 113 1 143 153 163 Axial levels 6-27 122 132 142 152 162 171 181 221 231 241 251 112 122 132 142 152 162 121 131 141 151 161 302 221 231 241 251 111 121 131 141 151 161 120 130 140 150 160 110 100 182 301 1 2 3 4 5 6 7 8 [P.J. Venter, M.N. Mitchell, F. Fortier, PBMR reactor design Core Radial Ring and development, in: Proceedings from the 18th Vessel core package nodalization Vessel control volume, flow path, and heat structure International Conference on Structural Mechanics in Reactor Technology (SMiRT 18), Beijing, China, Aug. 2005]

(8 rings x 29 axial levels) nodalization with core package boundaries in blue Correct aspect ratio 53

Reactor building Nodalization guided by NGNP layout

• Passive air-flow Reactor Cavity Cooling System (RCCS)

Leakage assumed to be the same as BWR Mark I reactor building surrounding the containment

• 100% vol/day at 0.25 psig HTGR Mechanistic Source Terms White Paper, July 2010, [INL-EXT-10-17999]

Picture above shows a water-cooled RCCS but demo model uses air-cooled RCCS. 54

Recirculation loop and secondary heat removal Recirculation system and secondary heat removal

• Recirculation loop and secondary heat removal provide boundary conditions to the vessel Vessel inlet Flow rate riser Cold leg Heat removal & inlet temperature 181 640 630 620 500 630 Pipe break nodalization allows Primary heat Hot leg 500 counter-current natural circulation 627 exchanger Vessel outlet 605 615 flow 605 plenum 610 510 600 610 110 600 620

• MELCOR counter-current flow model used to represent adjacent stream drag 625 626 forces

• Geometry similar to PWR hot leg natural circulation [NUREG-1922]

• Allows for air ingression Scenario: depressurized loss of forced circulation (DLOFC)

• Assumes double-ended break of the hot leg 55

DLOFC scenario DLOFC is initiated after 900 days of operation

• Long-term fission product concentrations developed in TRISO and pebble

• 24 kg/yr graphite dust generation based on German AVR experience

• TRISO initialized with 10-5 failure fraction during the steady state Provisions for air ingression Reactor cavity cooling system (RCCS) is operational Individual sensitivity calculations to explore variations in the model response to uncertainty in input parameters 56

DLOFC reference case results (1/7) Vessel Pressure 10 9.3 MPa initial pressure Following pipe break 9 8

• Control rods insert to terminate fission 7

• The vessel depressurizes in seconds as the high-Pressure (MPa) 6 pressure helium escapes out both sides of the broken 5 pipe 4 3

• Peak velocity in the pebble bed is 45 m/s (normal flow 2

rate is 11-18 m/s) 1 0

0 0.2 0.4 0.6 0.8 1 Time (min)

Pipe Break Flowrates 0.1 0.08 Vessel side lower Vessel side upper SG side 0.06 Positive flow is 0.04 out of the reactor Counter-current flow established on the vessel side Flowrate (kg/s) 0.02 of the pipe break 0

-0.02

• Hot gases from the exit plenum escape on the top -0.04 side of the broken hot leg pipe and cooler gases enter Negative flow is

-0.06 into the reactor along the bottom of the pipe -0.08

-0.1 0 24 48 72 Time (hr) 57

DLOFC reference case results (2/7) 0.100 Axial flow velocities in outer 0.075 Positive flow is region of the core In-vessel natural circulation flow after blowdown downward 0.050

• Upward flow in the inner region of the core where the fuel temperatures and decay power heating are 0.025 Velocity (m/s) higher 0.000

• Downward flow in the outer region of the core where -0.025 Negative flow is the fuel temperatures and decay power heating are -0.050 upward lower Axial flow velocities in inner

-0.075 region of the core

• Flow increases when the fuel starts to cool

-0.100 0 24 48 72 The fuel temperatures in the inner region of the 1800 Time (hr) pebble bed shift from cooler at inlet and hot at the 1600 outlet due to the flow reversal Exit hottest at the bottom 1400 of the pebble bed

• The axial fuel temperatures are affected by the local 1200 Temperature (deg-C) decay heat power (highest in the center) and the 1000 flow direction 800 During normal operation, the fuel at the exit (bottom) is 600 Exit is cooler due to the flow reversal and cooler air entering from the exit plenum the hottest 400 The exit becomes the coolest location (low power and 200 Top (inlet) is coolest when helium circulator is operating cooler gases entering from the exit plenum) 0 0 24 48 72 Time (hr) 58

DLOFC reference case results (3/7)

Decay Heat Power 5

The core heatup is dominated by the decay heat 4 Decay Heat

• The air oxidation power is relatively small at <25 kW Oxidation

• Although the vessel is thermally-stratified with a low 3 exit path, a small natural circulation flow persists to Power (MW) 2 bring air into the vessel Pebble bed inlet and circulation velocities are <0.04 m/s 1 Maximum oxidation power is <25 kW 0

0 24 48 72 96 120 144 168 Time (hr)

Graphite Oxdidation By-products The graphite oxidation produces significant 800 quantities of CO and CO2 700 CO 600

• Approximately 50% of the oxidation occurs in the CO2 500 graphite reflector structures around the inlet plenum and 50% in the lower portion of the pebble bed. Mass (kg) 400 300

• ~1% of the pebble matrix oxidized after 168 hr 200 17% peak pebble oxidation at the bottom center 100 0

0 24 48 72 Time (hr) 59

DLOFC reference case results (4/7)

CO Reactor Building Mole Fraction 1

Potential for combustion in the reactor building

• MELCOR lower limit for CO combustion with an ignition 0.8 source is 12.9% (~2X higher than for hydrogen)

Mole Fraction (-)

• Highly dependent on local concentrations and building 0.6 design and interconnectivity 0.4

• Demo reactor building assumes high inter-connectivity Lower CO Reactor building Allows air and CO circulation 0.2 flammability limit CO concentration

• No carbon-dioxide burns were predicted through 168 hr 0

0 24 48 72 Time (hr) 60

DLOFC reference case results (5/7)

MELCOR predicts release and transport from fuel to the environment

• Fuel heat-up

• TRISO layers - Initial failure fraction + failures during heat-up

• Pebble matrix and pebble outer shell - Higher diffusivity at elevated temperatures, recoil, and air oxidation

• Primary system - Failed with the initiating event

• Reactor building - Design leakage TRISO Failure Fraction vs Temperature TRISO Failure Fraction 1 1 0.1 0.1 Specified failure versus temperature Failure Fraction (-) Failure Fraction (-)

0.01 0.01

~2x10-4 failure fraction 0.001 calculated in the reference case 0.001 0.0001 0.0001 0.00001 0.00001 1200 1400 1600 1800 2000 2200 2400 2600 0 24 48 72 96 120 144 168 Temperature (°C) Time (hr) 61

DLOFC reference case results (6/7) Iodine Distribution 1.E-06 Release In-vessel The impact of the low TRISO failure 1.E-07 Reactor building Environment Fraction of initial inventory (-)

fraction leads to small releases 1.E-08

• Iodine diffusivity assumed to be same as 1.E-09 krypton 1.E-10 Initial release dominated

• Assumes most iodine reacts with cesium 1.E-11 by airborne dust from steady operation

• Larger cesium release due its the higher 1.E-12 0 24 48 72 diffusivity 1.E-03 Time (hr)

Cesium Distribution

• Ag release to the environment is 1.2x10-3 1.E-04 (highest diffusivity) 1.E-05 Fraction of initial inventory (-)

Release 1.E-06 In-vessel Reactor building 1.E-07 Environment 1.E-08 1.E-09 Initial release dominated 1.E-10 by airborne dust from steady operation 1.E-11 0 24 48 72 Time (hr) 62

DLOFC reference case results (7/7)

Iodine Distribution at 7 days 8.9%

Of the small release from the fuel 33.6%

34% and 62% of iodine and cesium, respectively, retained in the vessel

• Thermally-stratified orientation limits vessel releases

• Low flowrate combined with aerosol deposition 57.5%

• Inclusion of graphite oxidation reaction products (CO and CO2) promotes more flow and therefore more In-vessel Reactor building Environment releases from the vessel Cesium Distribution at 7 days 3.9%

58% and 34% of iodine and cesium, 34.4%

respectively, retained in the reactor building

• No strong driving force for reactor building leakage Reference model uses a hole size equivalent to 100%

leakage per day at a design pressure of 0.25 psig (3.2 in2) 61.7%

In-vessel Reactor building Environment 63

MELCOR can be used to explore the variability of the results to uncertainties Model Parameter Distribution Range Initial TRISO Failure Fraction (fraction of inventory) Log uniform 10 10-3 TRISO Failure Rate Multiplier (-) Log uniform 0.1 - 10.0 Intact TRISO Diffusivity Multiplier (-) Log uniform 0.001 - 1000.0 TRISO Model Failed TRISO Diffusivity Multiplier (-) Log uniform 0.001 - 1000.0 Parameters Matrix Diffusivity Multiplier (-) Log uniform 0.001 - 1000.0 TRISO Pebble Emissivity (-) Uniform 0.5 - 0.999 TRISO Pebble Bed Porosity (-) Uniform 0.3 - 0.5 TRISO recoil fraction (-) Uniform 0 - 0.03 Radionuclide Shape Factor (-) Uniform 1.0 - 5.0 Model Parameters Gaseous Iodine Multiplier (Base = 5% I2) Uniform 0.02 - 1.0 Graphite Conductivity Multiplier (-) Uniform 0.5 - 1.5 Decay Heat Multiplier (-) Uniform 0.9 - 1.1 RCCS Blockage Multiplier (-) Log uniform 0.001 - 1.0 Design Parameters RCCS Emissivity (-) Uniform 0.1 - 1.0 Reactor Building Leakage Multiplier (-) Log uniform 0.1 - 100.0 Wind speed (m/s) Uniform 0 - 10 64

Single parameter sensitivity results (1/4)

The sensitivity parameters were sampled at the minimum and maximum values to illustrate their impacts Peak Fuel Temperature 2000

• A low graphite conductivity has the largest impact on 1800 the peak fuel temperature 1600 Graphite conductivity varies considerably with irradiation

(>10X) and also varies with temperature 1400 Temperature (°C)

Base case 1200

• +/-10% decay heat has next largest impact on the peak Low Gr k High Gr k fuel temperature 1000 1.1X decay heat 0.9X decay heat 800

• High/low emissivity, the next most important single Low pebble emissivity High pebble emissivity factor, is used as a surrogate for the relative 600 Low pebble porosity High pebble porosity importance of radiative exchange in the pebble bed 400 Blocked RCCS

• Debris bed porosity had a small effect on the peak fuel 200 0 24 48 72 96 120 144 168 temperature Time (hr)

• Heat dissipation limits the magnitude of the initial peak for a blocked RCCS Slow heat-up to 1800 by 7 days 65

Examples of single parameter sensitivity results (2/4) 0.1 Base case TRISO Failure Fraction Low Gr k High Gr k 1.1X decay heat 0.01 0.9X decay heat Low pebble emissivity As the peak fuel temperature rises, the TRISO Failure Fraction (-)

High pebble emissivity Low pebble porosity failure fraction increases 0.001 High pebble porosity Blocked RCCS

• Blocked RCCS does not have impact for several days 0.0001 0.00001 0 24 48 72 96 120 144 168 Time (hr)

Cesium Release Fraction from the Pebbles 1.E+00 The cesium environmental release shows an 1.E-01 order of magnitude variation 1.E-02

• Reflects variations in release from the pebbles 1.E-03 Release Fraction (-)

• Graphite conductivity had the largest impact 1.E-04 Base case Low Gr k 1.E-05 High Gr k

• Variations in emissivity = uncertainty in radiative heat 1.E-06 1.1X decay heat 0.9X decay heat Low pebble emissivity transport (similar to +/-10% in decay heat power) 1.E-07 High pebble emissivity Low pebble porosity High pebble porosity

• Pebble porosity had a small impact 1.E-08 2x recoil No recoil Blocked RCCS 1.E-09 0 24 48 72 96 120 144 168 Time (hr) 66

Examples of single parameter sensitivity results (3/4) 1.E+00 Cesium Environmental Release 1.E-01 Larger hole size in the building and higher wind 1.E-02 speed causes higher releases to environment 1.E-03 Release Fraction (-)

• 100X building leakage has less than a 10X impact 1.E-04 1.E-05

• External wind has small effect 1.E-06 Base 1X RB Leakage, 10 m/s 1.E-07 10X RB leakage, 0 m/s wind Graphite oxidation and the associated CO/CO2 1.E-08 10X RB leakage, 10 m/s wind 100X RB leakage, 0 m/s wind production did not increase the source term 1.E-09 100X RB leakage, 10 m/s wind

• CO/CO2 gas production did not increase environment 1.E-10 0 24 48 72 96 120 144 168 release 1.E+00 Time (hr)

Cesium Environmental Release Early impacts of the recoil and initial TRISO failure 1.E-01 1.E-02 fraction did not impact long-term environmental 1.E-03 release Release Fraction (-)

1.E-04

• Magnitude of the release dominated by the fuel 1.E-05 Base case temperature response and the TRISO failure model 1.E-06 No oxidation 2X Recoil 1.E-07 No Recoil Late step change in the blocked RCCS release is 1.E-08 1.E-09 10X Initial TRISO failure 0.1X Initial TRISO failure Blocked RCCS due to a carbon monoxide burn 1.E-10 0 24 48 72 96 120 144 168

• Building pressurization forces out airborne radionuclides Time (hr) 67

Examples of single parameter sensitivity results (4/4) 1750 Graphite Oxdidation By-products Blocked RCCS leads to higher CO generation 1500 CO CO - Blocked RCCS CO2 CO2 - Blocked RCCS

• Ratio of reaction products is dependent on the 1250 temperature of the graphite 1000

• Blocked RCCS generates ~9% more moles of CO and Mass (kg) 750 CO2 500 Higher CO generation led to a burn in the steam 250 generator compartment (pipe break location) 0 0 24 48 72 96 120 144 168 Time (hr)

• Incomplete burn with slow flame speed Reactor Building CO Mole Fraction Low oxygen concentration (6.8%) 0.5 Base

• 0.25 bar (3.5 psi) pressure rise 0.4 Blocked RCCS CO Lower Flammability Limit

• Burn creates non-condensable CO2 Mole Fraction (-)

0.3 No subsequent condensation CO burn in the SG 0.2 compartment 0.1 0

0 24 48 72 96 120 144 168 Time (hr) 68

High-Temperature Gas-Cooled Reactor Uncertainty Analysis

Role of MELCOR in Resolving Uncertainty Uncertainty Engineering Performa nce Event Scenario Phenomenological Uncertainty Model Uncertainty Plant Ris k- Informed As s es s ment Initial/Boundary SSC Failure Modes Condition Uncertainty Simulation Uncertainty

Evolution from MELCOR LWR Uncertainty Analysis Overall motivation

• A clustering of system responses provides insights on important assumptions and modeling parameters

• Provides a most likely release and range of releases for the scenario MELCOR application to LWRs

• Range of SOARCA uncertainty studies

• PWR and BWR plant uncertainty studies

• Resolved role of uncertainty in critical severe accident issues Commonalities between LWR and HTGR

• Chemical form of key elements

• Aerosol physics parameters (e.g., shape factor)

• Operating time before accident happens

• Containment leakage hole size Parameter selection emphasized potential HTGR-specific uncertainties

• Ran 2000 realizations on High Performance Computer 71

Parametric Uncertainty - Capability Demonstration Model Parameter Distribution Range Initial TRISO Failure Fraction (fraction of inventory) Log uniform 10 10-3 TRISO Failure Rate Multiplier (-) Log uniform 0.1 - 100.0 Intact TRISO Diffusivity Multiplier (-) Log uniform 0.001 - 1000.0 TRISO Model Failed TRISO Diffusivity Multiplier (-) Log uniform 0.001 - 1000.0 Parameters Matrix Diffusivity Multiplier (-) Log uniform 0.001 - 1000.0 TRISO Pebble Emissivity (-) Uniform 0.5 - 0.999 TRISO Pebble Bed Porosity (-) Uniform 0.3 - 0.5 TRISO recoil fraction (-) Uniform 0 - 0.03 Radionuclide Shape Factor (-) Uniform 1.0 - 5.0 Model Parameters Gaseous Iodine Multiplier (Base = 5% I2) Uniform 0.02 - 1.0 Graphite Conductivity Multiplier (-) Uniform 0.5 - 1.5 Decay Heat Multiplier (-) Uniform 0.9 - 1.1 RCCS Blockage Multiplier (-) Log uniform 0.001 - 1.0 Design Parameters RCCS Emissivity (-) Uniform 0.1 - 1.0 Reactor Building Leakage Multiplier (-) Log uniform 0.1 - 100.0 Wind speed (m/s) Uniform 0 - 10 72

UO 2 Thermal Response 73

UO 2 Thermal Transient Evolution

• Core cells with peak fuel temperatures at end of simulation

• Simulation time denoted as accident phase

• These core cells do not exhibit cooldown prior to end of accident phase 74

TRISO Particle Failure Long-term TRISO particle failure possible for core cells exhibiting prolonged over-temperatures Initial distribution of failed TRISO particles 75

Evolution of TRISO Particle Failures Tails of realizations contributing Long-term failures of to longer term growth of TRISO TRISO particles at lower particle failures rate but driven by prolonged period of elevated fuel temperature 50th percentile reasonably stable in the long-term Lower rates of failure entirely driven by early temperature Rapid growth in failure excursion fraction driven by the Variability in peak fuel early temperature temperature and cooldown excursion transient dominates higher failure rate realizations 76

Role of Decay Heat Rejection - Latest Time to Peak Fuel Temperature 77

Role of Decay Heat Rejection - Peak Fuel Temperature 78

Summary Conclusions Added HTGR modeling capabilities to SCALE & MELCOR for HTGR source term analysis to show code readiness Modeling demonstrated for a DLOFC Scenario

• Input of detailed ORIGEN radionuclide inventory data from ORNL

• Input radial and axial power distributions from ORNL neutronic analysis

• Develop MELCOR input model for exploratory analysis

• Fast-running calculations facilitate sensitivity evaluations Developed an understanding of non-LWR beyond-design-basis-accident behavior and overall plant response 80

SCALE Overview SCALE Development for Regulatory Applications What Is It?

The SCALE code system is a modeling and simulation suite for nuclear safety analysis and design. It is a modernized code with a long history of application in the regulatory process.

How Is It Used?

SCALE is used to support licensing activities in NRR (e.g., analysis of spent fuel pool criticality, generating nuclear physics and decay heat parameters for design basis accident analysis) and NMSS (e.g., review of consolidated interim storage facilities, burnup credit).

Who Uses It?

SCALE is used by the U.S. Nuclear Regulatory Commission (NRC) and in 61 countries (about 10,000 users and 33 regulatory bodies).

How Has It Been Assessed?

SCALE has been validated against criticality benchmarks (>1000), destructive assay of fuel and decay heat for PWRs and BWRs (>200) 82

Data to generate for MELCOR: QOIs 83

VSOP Backup Slides VSOP workflow shares several features of conventional 2-step LWR core analyses Updated material inventories Core inventories Fuel shuffling /

pebble recycle Depletion update (pin / pebble)

Single-element /

assembly flux solution Few-group cross-sections Region-wise (critical spectrum) flux solution Simplified transport /

diffusion Spatial flux / power

• Homogenized material regions distribution across with few-group cross-sections the core 85

VSOP calculation flow (MEDUL) Fresh fuel pebbles Loaded material Ex-core decay Disposal Core material regions region storage storage Neutron spectrum Discharged batch YES Neutron diffusion (2-D / 3-D) NO BU BUmax ?

Burnup / depletion NO Discharged region YES Material update Thermal hydraulics keff ktarget ? (shuffle) 86

VSOP calculation flow (MEDUL) Fresh fuel pebbles Loaded material Ex-core decay Disposal Core material regions region storage storage Neutron spectrum Discharged batch YES This is just two-Neutron diffusion (2-D / 3-D) step neutronics NO (Polaris+PARCS) BU BUmax ?

Burnup / depletion NO Discharged region YES Material update Thermal hydraulics keff ktarget ? (shuffle) 87

VSOP calculation flow (MEDUL) Fresh fuel pebbles Loaded material Ex-core decay Disposal Core material regions region storage storage Neutron spectrum Discharged batch At equilibrium, spatial distributions YES are static: power, neutron spectrum, isotopics! NO Neutron diffusion (2-D / 3-D)

We use Monte Carlo to generate the BU BUmax ?

high-fidelity spatial flux spectrum and one-group cross sections Burnup / depletion NO Discharged region YES Material update Thermal hydraulics keff ktarget ? (shuffle) 88

VSOP calculation flow (MEDUL) Fresh fuel pebbles Loaded material Ex-core decay Disposal Core material regions region storage storage Neutron spectrum We simulate a pebble* moving Discharged batch through the equilibrium core with YES a time-dependent power and flux spectrum based on its position. NO Neutron diffusion (2-D / 3-D)

This pebble* can be used to BU BUmax ?

reconstruct the detailed core composition or iterate on the Burnup / depletion equilibrium core. NO Discharged region YES Material update Thermal hydraulics keff ktarget ? (shuffle)

• equivalent to a batch of pebbles with same history 89

Iterative procedure for developing equilibrium core compositions Determine average burnup of each pebble batch within a zone (axial / radial)

Deplete each batch within zone to its Repeat on initial guess respective burnup inventories until keff

• Origen library based on region-wise flux from core transport converges; depleted compositions represent approximate equilibrium Average zone compositions

• Weighted sum of batches Calculate core power distribution & flux shape by zone

• Generate ORIGEN library for each zone 90

Why use an iterative approach to equilibrium core compositions (instead of 2-step?)

• Were interested in determining equilibrium compositions and flux shape by region

• Not trying to perform dynamics or reload analysis; just need equilibrium in-core inventories

• At-equilibrium assumption simplifies analysis

• Conservative and bounding: i.e., converged upon highest core-averaged burnup (and thus highest fission product inventories)

• 2-step analysis requires many repeated calculations e.g., 22 axial zones x 5 passes through core => 110 calculations to perform one complete cycle! (Still not at equilibrium)

Feasible with few-group diffusion, costly for MG transport!

91

ORIGEN Library Interpolation Backup Slides

Aspects of the ORNL methodology for fuel inventory

• Rapid answers to common questions

• Up-front work required such as

• Sensitivity analysis of the reactor system What I/Cs/Pu content could I expect in a PBMR-400 to understand the state changes that pebble at 90 GWd/MTU? impact neutron flux spectrum in the fuel

a. assuming constant power? (e.g. moderator density in BWR)

b. pass-dependent power?

• Running many CPU-hours of TRITON

c. during a power maneuver? coupled transport+depletion cases to

d. after 4 days of decay? generate a database of 1-group cross

e. after 40 days of decay? sections which can be interpolated to a specific state (ORIGEN reactor library)

f. after 40 years of decay?

g. at 80 GWd/MTU?

• Those libraries can then be used later (in ORIGAMI) to regenerate inventory and

h. in a pebble with +1% enrichment? reaction rates:

Each answer requires a <10 second calc. on a single CPU () = () () ()

Why do it this way?

Why is speed important? This approach is not just for If is insensitive to decay time, power seeding MELCOR nodalizations. All back-end analysis level, then b through h can be answered can use this approach: dry storage casks, on-site storage, from a single TRITON pre-calculation!

discharge inventory analysis, transportation packages.

93

Strategy for LWRs

• What level of TRITON model fidelity is required to generate a reasonable 1-group xs database (ORIGEN reactor library) for rapid LWR inventory calculations?

Increasing fidelity

a. 3D full-core with plant-specific loading pattern Requires plant-specific knowledge

b. 3D full-core with equilibrium loading pattern Assembly position matters

c. 3D core subset Imposes additional assumptions

d. 3D single assembly or requires too much information!

e. 2D core subset

f. 2D single assembly Has trouble with local variations

g. 2D single pin (control elements, water holes, channel box)

h. 0D infinitely homogeneous mixture Has trouble if any geometry is important

• For LWRs, using 2D single assembly models to generate the 1-group xs database appears sufficient!

• verification confirms ORIGAMI reproduces TRITON results with same (simple) operating history

• validation against spent fuel inventory and decay heat measurements confirms the overall approach is adequate

• code results generally within experimental uncertainty bands

• <1% error in decay heat, <5% error in important nuclides, <15% error in others 94

Strategy for HTGRs

• What level of TRITON model fidelity is required to generate a reasonable 1-group xs database for rapid HTGR inventory calculations?

a. 3D full-core with plant-specific pebble Requires plant-specific knowledge loading & discharge strategy

b. 3D full-core with equilibrium pebble distribution Computationally expensive

c. 2D core slice with equilibrium pebble distribution Previously investigated in other work;

d. 1D single pebble with buffer for neighbor effects difficult to optimize buffer

e. 1D single pebble Does not account for reflectors

f. 0D infinitely homogeneous mixture Used in this study to understand sensitivity to model fidelity

• Using at SCALE/TRITON 3D full-core at equilibrium (b) is equivalent to VSOP but with:

• ENDF/B-VII.1+ modern nuclear data

• SCALE complete ORIGEN nuclide set instead of VSOP limited set

• SCALE high-fidelity full-core Monte Carlo transport instead of VSOP diffusion 95

Our focus for the PBMR-400

• First, understand the state changes that influence the neutron flux spectrum in a pebble as it flows through an equilibrium core:

a. pebble power history

b. pebble burnup

c. axial position in the core

d. radial position in the core (proximity to radial reflector)

e. pebble neighbors (burnup/temperature/inventory)

f. temperature

• Next, generalize the SCALE concept of the ORIGEN reactor library for HTGR / PBMR-400 96

Prototype ORIGAMI input for multi-pass pebble inventory calculations (SCALE 7.0)

ORIGAMI operating history input radial power shape pr = [ pr1 pr2 ... pr ]

axial power shape pz = [ p1 p2 pn ]

(relative) residence time in each axial zone ztime = [ rt1 rt2 rtn ]

Example history: 3-pass pebble history, each pass hist[

moves through declared axial zones pass{ power=180 burn=64 down=7 rzone=ANY }

pass{ power=160 burn=62 down=6 rzone=ANY }

power: average MWd/MTU for that pass pass{ power=140 burn=64 down=7 rzone=3 }

burn: days at power down: days decay ]

rzone: radial zone 97

Enhancing ORIGEN library interpolation capabilities to accommodate non-LWR systems

• Legacy ORIGEN library interpolation (via ARP) optimized for LWR analysis

• Interpolation dimensions of initial enrichment, average moderator density, burnup

• Diverse physics characteristics of non-LWR cores require new dimensions for reactor library interpolation

• e.g., PBMR: radial distance from reflector, initial pebble enrichment, reflector temperature

• To address this, we have developed a new HDF5-based format for self-describing ORIGEN libraries capable of accommodating arbitrary dimensions for interpolation 98

Legacy ORIGEN reactor data library interpolation relies on an ASCII database with hard-coded interpolation dimensions Pre-defined dimensions:

arpdata.txt

• 235U Enrichment

• Moderator density Assembly1 Assembly2 Assembly3 lib1 lib4 lib1 lib4 lib1 lib4 Individual permutations on lib2 lib5 lib2 lib5 lib2 lib5 interpolation dimensions lib3 lib6 lib3 lib6 lib3 lib6 99 Fuel Cycle Scenario Modeling Workshop - CyBORG 99

New HDF5-based Archive format designed to accommodate arbitrary interpolation dimensions HDF5 Archive Energy per Loss XS fission DecayData TransitionStructure Fission XS Energy per capture particle neutron Neutron yields Transition matrix Lib #1 Tags Initial enrichment Lib #2 Tags Refueling rate FP removal rate Lib #3 Tags 100

MELCOR for Accident Progression and Source Term Analysis

MELCOR Development for Regulatory Applications What Is It?

MELCOR is an engineering-level code that simulates the response of the reactor core, primary coolant system, containment, and surrounding buildings to a severe accident.

Who Uses It?

MELCOR is used by domestic universities and national laboratories, and international organizations in around 30 countries. It is distributed as part of NRCs Cooperative Severe Accident Research Program (CSARP).

How Is It Used?

MELCOR is used to support severe accident and source term activities at NRC, including the development of regulatory source terms for LWRs, analysis of success criteria for probabilistic risk assessment models, site risk studies, and forensic analysis of the Fukushima accident.

How Has It Been Assessed?

MELCOR has been validated against numerous international standard problems, benchmarks, separate effects (e.g., VERCORS) and integral experiments (e.g., Phebus FPT), and reactor accidents (e.g., TMI-2, Fukushima).

102

Source Term Development Process Experimental Basis PIRT process Oxidation/Gas Generation Melt Progression Fission Product Release Fission Product Transport Accident Analysis Design Synthesize MELCOR Scenario # 1 Scenario # 2 timings and Basis

. . release Source fractions Term Scenario # n-1 Scenario # n Cs Diffusivity 103

SCALE/MELCOR/MACCS Neutronics Integrated Severe Radiological SCALE MACCS MELCOR

• Criticality Accident Progression Consequences

• Shielding

• Hydrodynamics for range

• Near- and far-field

• Radionuclide inventory of working fluids atmospheric transport

• Burnup credit

• Accident response of and deposition

• Decay heat plant structures, systems

• Assessment of health and components and economic impacts

• Fission product transport Nuclear Reactor System Applications Non-Reactor Applications Safety/Risk Assessment Regulatory Design/Operational Support Fusion Spent Fuel Facility Safety

• Technology-neutral

• License amendments

• Design analysis scoping

• Neutron beam injectors

• Risk studies

• Leak path factor o Experimental

• Risk-informed regulation calculations

• Li loop LOFA transient

• Multi-unit accidents calculations o Naval

• Design certification (e.g.,

• Training simulators analysis

• Dry storage

• DOE safety toolbox codes o Advanced LWRs NuScale)

• ITER cryostat modeling

• Spent fuel

• DOE nuclear facilities o Advanced Non-LWRs

• Vulnerability studies

• He-cooled pebble test transport/package (Pantex, Hanford, Los

• Emergency preparedness blanket (H3) applications Alamos, Savannah River

• Accident forensics Site)

(Fukushima, TMI)

• Emergency Planning Zone

• Probabilistic risk Analysis assessment 104

MELCOR Attributes Foundations of MELCOR Development Fully integrated, engineering-level code Phenomena modeled

• Thermal-hydraulic response of reactor coolant system, reactor cavity, rector enclosures, and auxiliary buildings

• Core heat-up, degradation and relocation

• Core-concrete interaction

• Flammable gas production, transport and combustion

• Fission product release and transport behavior Level of physics modeling consistent with

• State-of-knowledge

• Necessity to capture global plant response

• Reduced-order and correlation-based modeling often most valuable to link plant physical conditions to evolution of severe accident and fission product release/transport Traditional application

• Models constructed by user from basic components (control volumes, flow paths and heat structures)

• Demonstrated adaptability to new reactor designs - HPR, HTGR, SMR, MSR, ATR, Naval Reactors, VVER, SFP, 105

MELCOR Attributes MELCOR Pedigree International Collaboration Cooperative Severe Accident Research Program (CSARP) - June/U.S.A Validated physical models MELCOR Code Assessment Program (MCAP) - June/U.S.A European MELCOR User Group (EMUG) Meeting - Spring/Europe

• International Standard Problems, benchmarks, experiments, and reactor European MELCOR User Group (EMUG) Meeting - Fall/Asia accidents

• Beyond design basis validation will always be limited by model uncertainty that arises when extrapolated to reactor-scale Cooperative Severe Accident Research Program (CSARP) is an NRC-sponsored international, collaborative community supporting the validation of MELCOR International LWR fleet relies on safety assessments performed with the MELCOR code 106

Common Phenomenology 107

MELCOR Modeling Approach Modeling is mechanistic consistent with level of knowledge of phenomena supported by experiments Parametric models enable uncertainties to be characterized

• Majority of modeling parameters can be varied

• Properties of materials, correlation coefficients, numerical controls/tolerances, etc.

Code models are general and flexible

• Relatively easy to model novel designs

• All-purpose thermal hydraulic and aerosol transport code 108

MELCOR State-of-the-Art MELCOR Code Development Version Date 2.2.18180 M2x Official Code Releases December 2020 2.2.14959 October 2019 2.2.11932 November 2018 2.2.9541 February 2017 2.1.6342 October 2014 2.1.4803 September 2012 2.1.3649 November 2011 2.1.3096 August 2011 2.1.YT August 2008 2.0 (beta) Sept 2006

MELCOR Software Quality Assurance - Best Practices MELCOR SQA Standards Emphasis is on Automation SNL Corporate procedure IM100.3.5 Affordable solutions CMMI-4+

NRC NUREG/BR-0167 Consistent solutions MELCOR Wiki Bug tracking and reporting

• Archiving information

• Bugzilla online

• Sharing resources (policies, conventions, information, progress) Code Validation among the development team.

• Assessment calculations

• Code cross walks for complex phenomena where Code Configuration Management (CM) data does not exist.

• Subversion

• TortoiseSVN Documentation

• Available on Subversion repository with links from

• VisualSVN integrates with Visual Studio wiki (IDE)

• Latest PDF with bookmarks automatically generated from word documents under Subversion Reviews control

• Code Reviews: Code Collaborator

• Links on MELCOR wiki

• Internal SQA reviews Project Management Continuous builds & testing

• Jira for tracking progress/issues

• DEF application used to launch multiple

• Can be viewable externally by stakeholders jobs and collect results

• Regression test report Sharing of information with users

• External web page

• More thorough testing for code release

• MELCOR workshops

• Target bug fixes and new models for

• MELCOR User Groups (EMUG & AMUG) testing 110

MELCOR Verification & Validation Basis LWR & non-LWR applications Volume 1: Primer & User Guide Volume 2: Reference Manual Volume 3: MELCOR Assessment Problems

[SAND2015-6693 R]

Specific to non-Analytical Problems Saturated Liquid Depressurization Adiabatic Expansion of Hydrogen AB-1 LOF,LOHS,TOP MSRE Air-Ingress Transient Heat Flow in a Semi-Infinite Heat Slab AB-5 TREAT M-Series LWR application experiments Helical SG HT T-3 ANL-ART-38 Cooling of Heat Structures in a Fluid Radial Heat Conduction in Annular Structures Establishment of Flow Sodium Fires Molten Salt Sodium Reactors HTGR (Completed) (planned) (planned) (planned) 111

Sample Validation Cases TRISO Diffusion Release Turbulent LACE LA1 and LA3 IAEA CRP-6 Benchmark tests experimentally Deposition Fractional Release examined the Case 1a 1b 2a 2b 3a 3b transport and US/INL 0.467 1.0 0.026 0.996 1.32E-4 0.208 retention of US/GA 0.453 0.97 0.006 0.968 7.33E-3 1.00 aerosols through US/SNL 0.465 1.0 0.026 0.995 1.00E-4 0.208 pipes with high US/NRC 0.463 1.0 0.026 0.989 1.25E-4 0.207 speed flow France 0.472 1.0 0.028 0.995 6.59E-5 0.207 Korea 0.473 1.0 0.029 0.995 4.72E-4 0.210 Germany 0.456 1.0 0.026 0.991 1.15E-3 0.218 Resuspension (1a): Bare kernel (1200 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) A sensitivity study to examine STORM (Simplified Test of Resuspension (1b): Bare kernel (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) fission product release from Mechanism) test facility (2a): kernel+buffer+iPyC (1200 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) a fuel particle starting with a (2b): kernel+buffer+iPyC (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />)

(3a): Intact (1600 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) bare kernel and ending with (3b): Intact (1800 oC for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />) an irradiated TRISO particle; Aerosol Physics

• Agglomeration

• Deposition

• Condensation and Evaporation at surfaces Validation Cases

• Simple geometry: AHMED, ABCOVE (AB5 & AB6), LACE(LA4),

• Multi-compartment geometry: VANAM (M3), DEMONA(B3)

• Deposition: STORM, LACE(LA1, LA3) 112

MELCOR Modernization Generalized numerical solution engine Hydrodynamics In-vessel damage progression Ex-vessel damage progression Fission product release and transport 113

MELCOR default radionuclide classes

MELCOR default radionuclide classes 115